[en] At TPL (Tritium Process Laboratory) of JAEA, ITER relevant tritium technologies have been carried out. The design studies of Air Detirtiation System have been carried out in JAEA as a contribution of Japan to ITER. For the tritium processing technologies, our efforts have been focused on the R and D of the tritium recovery system of ITER test blanket system. A ceramic proton conductor has been studied as an advanced blanket system. It was observed that the hydrogen transportation rate of the conductor increased with the areas of the triple phase boundary (the contact point of gas, electrolyte and electrode). To decrease the grain size of the metal electrode by sputtering may be a solution for the increase of the areas of the triple phase. A series of fundamental studies on tritium safety technologies not only for ITER but also for fusion DEMO plants has also been carried out at TPL of JAEA. The main R and D activities in this field are the tritium behavior in a confinement and its barrier materials; monitoring; accountancy; detritiation and decontamination etc. The retention of hydrogen on a tungsten surface was studied by exposing a high flux and low energy deuterium plasma. As a new typical observation, a blister bursting with a tail, or a small hole or a vanished cap was found on some grains after the plasma exposure in spite of the low energy plasma. In this paper, the results of above recent activities at TPL of JAEA are summarized from viewpoint of ITER relevant and future fusion DEMO reactors. (orig.)

[en] Depth profiles of the Ag grains density in autoradiographs, which represent tritium concentration in CX-2002U samples exposed to high flux D/T particles under various conditions, were examined and the apparent diffusion coefficients were estimated from the profiles. Plasma discharge generating D/T atomized particles with low energies increases tritium inventory in the samples by introducing high tritium concentration on the surface exposed and following diffusion process into the deep region with apparent diffusion coefficients (1.7x10^-16 m²/s at 293 K and 2.3x10^-15 m²/s at 573 K), which are much larger than the diffusion coefficients in the bulk reported. Oxygen RF-plasma exposure might be effective to remove tritium retained even at a fairly deep region in carbon fiber composite (CFC) components

[en] Light amplification in a collective atomic recoil laser with an active medium confined in an external harmonic trap is investigated. Our study confirms the possibility of a great promotion of the lasing efficiency due to the collective atomic recoil motion being modified by the trapping potential. The physical explanation of this laser behavior is given

[en] The decomposition of methane over Zr, Zr₄Ni and Zr₂Ni was investigated to develop highly active materials for capturing tritiated methane inevitably formed in tritium handling systems. The entire decomposition or absorption curves, however, could not been described by any simple kinetic equations appearing in the literature. The present paper describes verification of plausible kinetic equations reproducing the observed absorption curves, assuming a reaction mechanism consisting of a progressive removal of hydrogen atoms according to CH₄(g)→CH₃(α)→CH₂(α)→CH(α)→C(α) and its modification including carbon segregation schemes by solving a set of kinetic equations by means of finite difference method, by taking into account of ab-initio calculations of potential energy surfaces by Gaussian 03. It was found that a step-by-step H-deletion model could not generate the experimentally observed absorption curves, but they could be reproduced quite well in the whole reaction range by a modified reaction scheme assuming coagulation of a carbon residue like CH₂(α) or CH(α) to carbonaceous deposits, described as CH₄(g)(k₁ decay)CH₃(α)(k₂ decay)CH(α)(k₃ decay)C-deposits. It was concluded that the final third step (with a rate constant of k₃) governs the overall absorption reactivity (methane consumption beyond 90%) of the materials. (author)

[en] The decomposition of methane and deutero-methane (CD₄) over Zr, Zr₄Ni and Zr₂Ni was investigated to develop highly active materials for capturing tritiated methane inevitably formed in tritium handling systems. None of the decomposition or absorption curves could been described by any simple kinetic equations appearing in the literature. They could be described by a series reaction scheme including growth of carbonaceous deposits on the surface. The reaction mechanisms were studied in more detail in the present study, and it was revealed that the absorption/decomposition of methane proceeds via the steps CH₄(g) (k₁ decay) CH₃(α) (k₂ decay) CH₂(α) (k₃ decay) C-deposits. The kinetic isotope effect on the decomposition/absorption of CH₄ and CD₄ was also analyzed for Zr in the light of ab initio calculations using small clusters including two or three Zr atoms. (author)

[en] A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm², and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed

[en] In flow measurement by means of transit time correlation or simplified methods the measurement system is regarded to be a linear time invariant system. In the present paper the general shape of the impulse-response is examined and the resulting systematic errors are discussed. The stochastic errors of the methods are taken into account, too. (orig.)

[en] The authors report on the structural and electrical characteristics of high-k erbium titanium oxide (Er₂TiO₅) gate dielectrics deposited on Si (100) substrates by reactive rf sputtering. They find that the capacitance value of Er₂TiO₅ gate dielectric annealed at 700 deg. C is higher compared to other annealing temperatures and exhibits a lower hysteresis voltage as well as interface trap density in C-V curves. This dielectric also shows almost negligible charge trapping under high constant voltage stress. This phenomenon is attributed to an amorphous Er₂TiO₅ structure and the suppression of the interfacial layer and Er silicate observed from x-ray diffraction and x-ray photoelectron spectroscopy, respectively

[en] A first engineering type experiment for evaluation of tritium permeation rate through a tungsten armored divertor of a DT fusion machine was carried out using tritium plasma experimental (TPE) apparatus under the collaborative program between US-DOE and JAERI. A test module having a multi-layer structure (W:1 mm^t, Cu:5 mm^t and a cavity containing water pressurized at 2.2 MPa) was exposed to DT (D:T=2:1) plasma. Break-through of tritium permeation into the water was observed after 4 h of plasma exposure. The amount of tritium permeated was compared with permeation rates predicted by numerical simulation codes (TMAP4 and TPERM). Feasibility of the experimental system and procedures to predict the tritium permeation behavior was examined using the simulation codes

[en] In fusion reactor, a few kg of tritium or more will be stored in hydride beds. In case of ITER, a ZrCo bed for storing 100 g of tritium has been designed with self-accountancy ability and more than 30 of beds will be used in the fuel storage and delivery system. In order to enhance total safety of fusion facility with tritium, the safe design and operation of storage beds will be one of the most important points. Concerning the safety design, the effect of tritium decay, such as decay heat transfer and He-3 behavior, is a key issue with the normal protection of over temperature, over pressure and leak for a metal-hydride bed. Concerning on the safety operation, procedure of hydrogenation-dehydrogenation cycle is a key issue under the requirements of the storage system in the fusion facility. Also, emergency performances, such as a rapid hydrogen recovery and loss of normal cooling function, should be investigated. In the Tritium Process Lab. in Japan Atomic Energy Agency, many tritium storage beds with ZrCo and U have been used with/without self-accountancy ability, and the safe handling experiences have been accumulated for almost 20 years. In this paper, the experiences concerning the above key issues and failures summarized, and the enhancement of safety is discussed for future tritium storage system. (orig.)